1. Field
Example embodiments relate to inspection, maintenance, and repair apparatuses and methods for nuclear reactors. Additionally, example embodiments relate to inspection, maintenance, and repair apparatuses and methods for nuclear reactors in confined areas, such as within the downcomer annulus between the reactor pressure vessel and the core shroud.
2. Description of Related Art
FIG. 1 is a sectional view, with parts cut away, of a typical reactor pressure vessel (“RPV”) 100 in a related art nuclear boiling water reactor (“BWR”). During operation of the BWR, coolant water circulating inside RPV 100 is heated by nuclear fission produced in core 102. Feedwater is admitted into RPV 100 via feedwater inlet 104 and feedwater sparger 106 (a ring-shaped pipe that includes apertures for circumferentially distributing the feedwater inside RPV 100). The feedwater from feedwater sparger 106 flows downwardly through downcomer annulus 108 (an annular region between RPV 100 and core shroud 110).
Core shroud 110 is a stainless steel cylinder that surrounds core 102. Core 102 includes a multiplicity of fuel bundle assemblies 112 (two 2×2 arrays, for example, are shown in FIG. 1). Each array of fuel bundle assemblies 112 is supported at its top by top guide 114 and at its bottom by core plate 116. Top guide 114 provides lateral support for the top of fuel bundle assemblies 112 and maintains correct fuel-channel spacing to permit control rod insertion.
The coolant water flows downward through downcomer annulus 108 and into core lower plenum 118. The coolant water in core lower plenum 118 in turn flows upward through core 102. The coolant water enters fuel bundle assemblies 112, wherein a boiling boundary layer is established. A mixture of water and steam exits core 102 and enters core upper plenum 120 under shroud head 122. Core upper plenum 120 provides standoff between the steam-water mixture exiting core 102 and entering standpipes 124. Standpipes 124 are disposed atop shroud head 122 and in fluid communication with core upper plenum 120.
The steam-water mixture flows through standpipes 124 and enters steam separators 126 (which may be, for example, of the axial-flow, centrifugal type). Steam separators 126 substantially separate the steam-water mixture into liquid water and steam. The separated liquid water mixes with feedwater in mixing plenum 128. This mixture then returns to core 102 via downcomer annulus 108. The separated steam passes through steam dryers 130 and enters steam dome 132. The dried steam is withdrawn from RPV 100 via steam outlet 134 for use in turbines and other equipment (not shown).
The BWR also includes a coolant recirculation system that provides the forced convection flow through core 102 necessary to attain the required power density. A portion of the water is sucked from the lower end of downcomer annulus 108 via recirculation water outlet 136 and forced by a centrifugal recirculation pump (not shown) into a plurality of jet pump assemblies 138 (only one of which is shown) via recirculation water inlets 140. The jet pump assemblies 138 are circumferentially distributed around the core shroud 110 and provide the required reactor core flow. A typical BWR includes 16 to 24 inlet mixers.
As shown in FIG. 1, related art jet pump assemblies 138 typically include a pair of inlet mixers 142. Each inlet mixer 142 has an elbow 144 welded thereto which receives pressurized driving water from a recirculation pump (not shown) via inlet riser 146. An exemplary inlet mixer 142 includes a set of five nozzles circumferentially distributed at equal angles about the inlet mixer axis. Each nozzle is tapered radially inwardly at its outlet. The jet pump is energized by these convergent nozzles. Five secondary inlet openings are radially outside of the nozzle exits. Therefore, as jets of water exit the nozzles, water from downcomer annulus 108 is drawn into inlet mixer 142 via the secondary inlet openings, where it is mixed with coolant water from the recirculation pump. The coolant water then flows into diffuser 148.
Core shroud 110 may include, for example, a shroud head flange (not shown) for supporting shroud head 122, an upper shroud wall (not shown) having a top end welded to the shroud head flange, a top guide support ring (not shown) welded to the bottom end of the upper shroud wall, a middle shroud wall (not shown) having a top end welded to the top guide support ring and including two or three vertically stacked shell sections (not shown) joined by mid-shroud attachment weld(s), and an annular core plate support ring (not shown) welded to the bottom end of the middle shroud wall and to the top end of a lower shroud wall (not shown). The entire shroud is supported by a shroud support (not shown), which is welded to the bottom of the lower shroud wall, and by an annular jet pump support plate (not shown), which is welded at its inner diameter to the shroud support and at its outer diameter to RPV 100.
Typically, the material of core shroud 110 and associated welds is austenitic stainless steel having reduced carbon content. The heat-affected zones of the shroud girth welds, including the mid-shroud attachment weld(s), have residual weld stresses. Therefore, mechanisms are present for mid-shroud attachment weld(s) and other girth welds to be susceptible to intergranular stress corrosion cracking (IGSCC).
IGSCC in the heat affected zone of any shroud girth seam weld diminishes the structural integrity of core shroud 110, which vertically and horizontally supports top guide 114 and shroud head 122. In particular, a cracked core shroud 110 increases the risks posed by a loss-of-coolant accident (LOCA) or seismic loads. During a LOCA, the loss of coolant from RPV 100 produces a loss of pressure above shroud head 122 and an increase in pressure inside core shroud 110, i.e., underneath shroud head 122. The result is an increased lifting force on shroud head 122 and on the upper portions of core shroud 110 to which shroud head 122 is bolted. If core shroud 110 has fully cracked girth welds, the lifting forces produced during a LOCA could cause core shroud 110 to separate along the areas of cracking, producing undesirable leaking of reactor coolant. Also, if the weld zones of core shroud 110 fail due to IGSCC, there is a risk of misalignment from seismic loads and damage to core 102 and the control rod components, which would adversely affect control rod insertion and safe shutdown.
Thus, core shroud 110 needs to be examined periodically to determine its structural integrity and the need for repair. Ultrasonic inspection is a known technique for detecting cracks in nuclear reactor components. The inspection area of primary interest is the outside surface of core shroud 110 at the horizontal and/or vertical mid-shroud attachment weld(s). However, core shroud 110 is difficult to access. Installation access is limited to the annular space between the outside of core shroud 110 and the inside of RPV 100, between adjacent jet pump assemblies 138. Scanning operation access is additionally restricted within the narrow space between core shroud 110 and jet pump assemblies 138, which is about 0.5 inch wide in some locations. The inspection areas are highly radioactive and may be located under water, 50 feet or more below an operator's work platform. As a result, inspection of core shroud 110 and/or RPV 100, as well as all other inspection, maintenance, and repair within downcomer annulus 108 often is difficult and complicated.
Solutions to the problem of inspecting core shroud 110 have been proposed, as discussed, for example, in U.S. Pat. No. 5,586,155 (“the '155 patent”). The disclosure of the '155 patent is incorporated in this application by reference. However, these proposed solutions do not include inspection, maintenance, and repair apparatuses and methods for nuclear reactors similar to the present invention.